1. Field of the Invention
The present invention relates to an underwater inspection/repair apparatus and, more particularly, an underwater inspection/repair apparatus capable of conducting inspection/repair of an interior of a reactor vessel without discharging a water from the reactor vessel.
2. Description of the Related Art
A boiling water reactor as one type of a light water reactor has a configuration shown in FIG. 6, for example. In FIG. 6, a reference 60 denotes a reactor. The reactor 60 comprises a reactor pressure vessel 62 having a top removable cover 61. A core 64 consisting of a plurality of fuel assemblies 63, 63, . . . , 63 is provided in the reactor pressure vessel 62. Each of the fuel assemblies 63 includes a plurality of elongate fuel rods (not shown). Each of fuel rods is constructed by covering a uranium dioxide pellet with a cladding tube. A steam separator 65 is provided over the core 64 and then a steam dryer 66 is provided over the steam separator 65.
A plurality of control rods 67, 67, . . . , 67 are inserted into clearances between the fuel assemblies 63, 63, . . . , 63 to be movable along their longitudinal direction. These control rods 67, 67, . . . , 67 can be driven vertically by a control rod drive mechanism (CRD) 68. The control rod drive mechanism 68 has rods 69,69, . . . , 69 connected to the control rods 67, 67, . . . , 67 respectively. These rods 69, 69, . . . , 69 are inserted respectively into cylindrical housings (through pressure-vessel housings) 70,70, . . . , 70 which extend into the inside of the reactor pressure vessel 62 via a bottom portion of the reactor pressure vessel 62. Flanges 71, 71, . . . , 71 whose diameters are set larger than outer diameters of the housings 70 are provided to lower end portions of these housings 70, 70, . . . , 70 to fit a main body of the control rod drive mechanism.
A substantially cylindrical core shroud 72 is provided around the core 64. A plurality of jet pumps 73, 73, . . . , 73 are provided in clearances between the core shroud 72 and an inner wall of the reactor pressure vessel 62. A recirculation water inlet nozzle 74 and a recirculation water outlet nozzle 75 are provided on a side peripheral wall of the reactor pressure vessel 62 to pass through the vessel wall. The recirculation water inlet nozzle 74 and the recirculation water outlet nozzle 75 are connected via a recirculation loop 76 provided on the outside of the reactor pressure vessel 62. One end of the recirculation loop 76 is positioned so as to oppose to a nozzle 73a of the jet pump 73 via the recirculation water inlet nozzle 74. A reactor recirculation pump 77 is interposed in the middle of the recirculation loop 76.
A main steam outlet nozzle 79 is provided on a side peripheral wall of the reactor pressure vessel 62 to pass through the vessel wall. A main steam pipe 81 is connected to the reactor pressure vessel 62. A through pressure-vessel nozzle 78 for measuring a water level is also provided on the side peripheral wall of the reactor pressure vessel 62 to pass through the vessel wall. FIG. 7 shows details around the through pressure-vessel nozzle 78. As can be seen from FIG. 7, a cladding portion 82 made of stainless steel is formed by welding on an inner wall surface of the reactor pressure vessel 62. A welded portion 83 made of inconel alloy which is excellent in both heat resistance and corrosion resistance is formed on the end portion of the through pressure-vessel nozzle 78 on the core 64 side.
An inside of the reactor pressure vessel 62 is filled with a core water (light water) W such that the core 64 is sufficiently covered with the water W. The core water W can function as moderator and coolant of the reactor 60.
As shown in FIG. 8, a fuel exchanger 84 which performs mainly exchange and replacement of the fuel assemblies 63 is provided over the reactor pressure vessel 62. When the fuel assemblies 63 are exchanged by using the fuel exchanger 84, the top removable cover 61 of the reactor pressure vessel 62 is removed.
In the boiling water reactor having the above configuration, heat can be generated by fission reaction of uranium in the fuel rods constituting the fuel assemblies 63 and then a core water W can be boiled by such heat. The boiled core water W can be separated into steam and water by virtue of the steam separator 65. Then, the separated steam can be dried by virtue of the steam dryer 66 and then supplied to a steam turbine (not shown) via the main steam outlet nozzle 79 and the main steam pipe 81. The steam, when supplied to the steam turbine, can drive the steam turbine. The steam can then be condensed by the condenser (not shown), and then can be circulated back into an inside of the reactor pressure vessel 62 via a water feed pipe (not shown) and a water feed nozzle (not shown).
Meanwhile, the core water W, when supplied to the nozzles 73a of the jet pumps 73 by the reactor recirculation pump 77, is pressurized downward by the jet pumps 73 to enter into the bottom portion of the core 64, and then the flow of the core water W is changed upward to flow into the inside of the core 64. The core water W can be circulated effectively by using the jet pumps 73 in this manner. The control rod drive mechanism 68 can insert and pull out the control rods 67, 67, . . . , 67 by moving the rods 69, 69, . . . , 69 vertically by means of hydraulic pressure drive, for example, so that it can control the output of the reactor 60 by absorbing neutrons emitted by nuclear fission.
However, for example, if austenitic stainless steels (e.g., SUS 304, etc.) are employed as material for the through pressure-vessel nozzle 78, there has been such a possibility that, under certain conditions, stress corrosion crackings (SCCs) occur in the welded portion between the through pressure-vessel nozzle 78 and the reactor pressure vessel 62 or in the through pressure-vessel nozzle 78 in vicinity of the welded portion.
Such stress corrosion crackings may be caused when three factors, i.e., sensitization of material (i.e., a phenomenon that a chromium depletion layer is generated in the neighborhood of grain boundary because of heat affection of the welding to thus degrade corrosion resistance), welding residual stress caused in the welded portion, and high temperature core water environment including a very small amount of dissolved oxygen are superposed.
Accordingly, the stress corrosion crackings can be prevented by reducing the degrees of the above three factors or eliminating more than one of above three factors, and therefore various countermeasures have already been taken. There have been possibilities that rust, crackings, etc. are generated in the inner surface of the through pressure-vessel nozzle 78, etc. due to any causes in addition to the above stress corrosion crackings.
In the related art, if crackings are generated in the through pressure-vessel nozzle 78, etc. because of the above stress corrosion crackings and other causes, the core water W filled in the reactor pressure vessel 62 has had to be discharged from the reactor pressure vessel 62 to carry out the repair operation. Then, after the core water W has been discharged, the operators have performed disconnection of the pipes, etc. from the outside of the reactor pressure vessel 62.
In this manner, since the related art repair operation has had to be conducted after the core water W filled in the reactor pressure vessel 62 has been discharged therefrom, not only longer hours have been required for a working time, but also a dose rate has been increased in the working environment because of loss of the radiation shielding effect obtained by the core water W. As a result, it has been extremely difficult to perform the repair operation quickly with regard to the permissible exposure doze for the operator.